Potential of Thorium Based Fuel Cycles to Constrain Plutonium and Reduce Long Lived Waste Toxicity

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IAEA-TECDOC-1349

Final report of a co-ordinated research project 1995–2001

April 2003

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The originating Section of this publication in the IAEA was:

Nuclear Power Technology Development Section International Atomic Energy Agency

Wagramer Strasse 5 P.O. Box 100 A-1400 Vienna, Austria

POTENTIAL OF THORIUM BASED FUEL CYCLES TO CONSTRAIN PLUTONIUM AND REDUCE LONG LIVED WASTE TOXICITY

IAEA, VIENNA, 2003 IAEA-TECDOC-1349 ISBN92–0–103203–X

ISSN 1011–4289

Printed by the IAEA in Austria

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FOREWORD

An important function of the International Atomic Energy Agency is to "foster the exchange of scientific and technical information" and to "encourage and assist research on, and development and practical application of, atomic energy for peaceful uses throughout the world". For innovative advanced nuclear reactor concepts, IAEA Member States in many cases find it attractive to co-operate internationally in technology development. The IAEA's fast reactor and hybrid systems technology development activities, which are conducted within its nuclear power programme, encourage international co-operation through technical information exchange and collaborative research. As regards the latter, co-ordinated research projects (CRPs) are tools that are effectively used in the implementation of the IAEA’s activities, both to promote exchange of scientific and technical information, and to pursue collaborative research and development tasks. Apart from allowing the efforts to be shared on an international basis and benefitting from the joint experience and expertise of researchers from the participating institutes, CRPs foster international team building.

From 1995 to 2001, the IAEA initiated a CRP on Potential of Thorium Based Fuel Cycles to Constrain Plutonium and to Reduce Long Term Waste Toxicity. The Member States involved in the CRP were: China, Germany, India, Israel, Japan, Republic of Korea, Netherlands, Russian Federation and the United States of America.

The research programme was divided into three stages: (1) benchmark calculations, (2) optimization of the incineration of plutonium in various reactor types, and (3) assessment of the resulting impact on the waste radio toxicity.

The results of all three stages were presented at international conferences, specifically, ICENES 98, ICENES 2000, and PHYSOR 2002 respectively.

The present report was prepared with the assistance of H.J. Rütten, Research Center Jülich (FZJ, Germany). The IAEA officer responsible for this publication was A. Stanculescu of the Division of Nuclear Power.

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EDITORIAL NOTE

In preparing this publication for press, staff of the IAEA have made up the pages from the original manuscript(s). The views expressed do not necessarily reflect those of the IAEA, the governments of the nominating Member States or the nominating organizations.

Throughout the text names of Member States are retained as they were when the text was compiled.

The use of particular designations of countries or territories does not imply any judgement by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries.

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CONTENTS

1. INTRODUCTION... 1

2. EXECUTIVE SUMMARY AND CONCLUSIONS ... 2

2.1. Comparison of methods and basic nuclear data... 2

2.1.1. Cell burnup calculations ... 2

2.1.2. Lattice calculations for LWR... 9

2.2. Evaluation of the potential of LWRs, HTRs, HWRs and MSRs for plutonium incineration... 14

2.2.1. Incentives... 14

2.2.2. Results... 15

2.2.3. Conclusions... 16

2.3. Effect of plutonium incineration on the toxicity of disposed nuclear waste ... 18

2.3.1. Incentives and database ... 18

2.3.2. Toxicity benchmark ... 18

2.3.3. Possible reduction of the radio-waste toxicity... 22

2.3.4. Results and conclusions... 23

2.4. Conclusions... 26

References to Section 2 ... 27

3. INDIVIDUAL CONTRIBUTIONS OF THE VARIOUS COUNTRIES... 28

3.1. China... 28

3.1.1. Study of thorium fuel cycles burning weapons grad and civil grade plutonium in the Module-HTR ... 28

3.1.2. Physics studies of energy production and plutonium burning in pebble-bed type high temperature gas cooled module reactor ... 32

References to Section 3.1... 35

3.2. Germany... 36

3.2.1. Introduction... 36

3.2.2. Optimization of plutonium incineration in the modular HTR ... 36

3.2.3. Effect of plutonium incineration on the long lived waste toxicity ... 45

3.2.4. Summary and conclusions ... 48

3.3. India ... 50

3.3.2. Benchmarks ... 50

3.3.3. Evaluation of the potential of HWRs for plutonium incineration ... 51

3.3.4. Assessment of the effect of plutonium incineration on waste toxicity... 52

3.3.5. Details of reactor calculations for plutonium burner (PHWR)... 61

3.4. Israel and the USA... 66

3.4.2. Toxicity calculations... 73

3.5. Japan ... 79

3.5.2. Reactor model... 79

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3.5.3. Calculation of fuel depletion ... 80

3.5.4. Calculation of toxicity ... 81

3.5.5. Conclusion ... 82

3.6. Republic of Korea... 92

3.6.1. Potential of a thorium based fuel cycle for 900 MW(e) PWR core to incinerate plutonium... 92

3.6.2. Assessment of the effect of plutonium incineration on the long lived waste toxicity... 101

3.7. Russian Federation... 106

3.7.1. Calculations on the principal neutronics characteristics of the WWER-1000 reactor loaded with PuO2–ThO2 fuel based on weapons grade plutonium... 106

3.7.2. Calculations of the principal neutronics characteristics of the WWER-1000 reactor loaded with PuO2–ThO2 fuel based on reactor grade plutonium ... 116

3.7.3. Assessment of the effect of plutonium burning on the waste toxicity... 120

3.8. Netherlands ... 123

3.8.2. Calculation method... 123

3.8.3. Results of the benchmark calculation ... 124

3.8.4. Numerical results of the benchmark ... 128

PARTICIPANTS IN THE CO-ORDINATED RESEARCH PROJECT... 133

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1. INTRODUCTION

Large stockpiles of civil plutonium have accumulated in the world from the different countries’ nuclear power programs. There is a serious public and political concern in the world about misuse of this plutonium and about accidental release of highly radiotoxic material into the environment. It therefore becomes necessary to keep the plutonium under strong security. One alternative for the management of plutonium is to incinerate it in reactors. But if the plutonium is fueled in reactors in the form of uranium/plutonium mixed oxide (MOX), second-generation plutonium is produced. A possible solution to this problem is to incinerate plutonium in combination with thorium. The thorium cycle produces ²³³U which, from a non-proliferation point of view, is preferable to plutonium for two reasons.

Firstly, it is contaminated with ²³²U, which decays to give highly active daughter products.

This would make handling and diversion difficult. Secondly, in case this is not sufficient deterrent, the ²³³U could be denatured by adding some U²³⁸ to the thorium. The quantity of

238U could be fine-tuned so as to be sufficient to denature the ²³³U, but not so much as to produce a significant quantity of plutonium. The thorium option not only produces electricity, but also replaces the plutonium with denatured ²³³U, which can be used in other reactors at a later date. All this can be done in existing reactors.

In the framework of IAEA activities on the use of thorium as nuclear fuel, a report on the performance of the thorium cycle, entitled A Fresh Look at the Thorium Fuel Cycle was drafted in 1991 and distributed as Working Material. IAEA-TECDOC-1155, entitled Thorium Based Fuel Options for the Generation of Electricity: Developments in the 1990s, was published as a follow-up action.

Co-ordinated Research Projects (CRPs) are tools that are effectively used by the IAEA to promote exchange of scientific and technical information and assist advanced nuclear power reactor technology research and development. CRPs allow the sharing of efforts on an international basis, benefiting from the experience and expertise of researchers from the participating institutes, and fostering international team building.

At the Consultants Meeting on Important Consideration on the Status of Thorium held in Vienna from 29 November to 1 December 1994, participants recommended the IAEA to organize a CRP on thorium-based fuel cycle issue. In 1995, the IAEA approved the topic for the CRP: Potential of Thorium based Fuel Cycles to Constrain Plutonium and to Reduce Long term Waste Toxicity. The scope of this CRP was discussed and agreed upon by the participants of the Consultants Meeting on Thorium based Fuel Cycles, held from 6 to 9 June 1995 at the IAEA in Vienna. The participating countries in the CRP were: China, Germany, India, Israel, Japan, Republic of Korea, Netherlands, Russian Federation and the United States of America.

This CRP examined the different fuel cycle options in which plutonium can be recycled with thorium to incinerate the burner. The potential of the thorium-matrix has been examined through computer simulations. Each participant has chosen his own cycle, and the different cycles were compared through certain predefined parameters (e.g., annual reduction of plutonium stockpiles). The toxicity accumulation and the transmutation potential of thorium-based cycles for current, advanced and innovative nuclear power reactors were investigated. As a final outcome, the CRP as next step would suggest to concentrate on the practical demonstration of plutonium-thorium incineration in a reactor in one of the member countries.

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2. SYNTHESIS OF THE CO-ORDINATED RESEARCH PROJECT’S TASKS AND RESULTS

2.1. COMPARISON OF METHODS AND BASIC NUCLEAR DATA 2.1.1. Cell-burnup calculations

In order to establish a comparison of the effect of different methods and databases applied in the countries participating in the CRP, benchmark calculations had to be performed before the start of the actual fuel cycles studies. For the first plutonium incineration benchmark calculations, the PWR-type reactor has been chosen because it is the reactor type that has the largest share in the current production of nuclear energy. The following topic was selected for the IAEA benchmark 1:

“Calculation of the isotopic composition, cross-sections and fluxes for a typical PWR-cell loaded with (Pu-Th)O2 - fuel, as a function of the fuel burnup.”

2.1.1.1. Definition of the fuel cell and tasks

The geometry of the reference fuel cell is displayed in Fig. 2.1. Table 2.1 gives the description of the material composition in terms of nuclide concentrations for the different cell zones.

Infinite lattice

Average power: P = 211 W/cm

Average temperature of the fuel: Tfuel = 1023 K Average temperature of the water: Tmod = 583 K

FIG. 2.1. Layout of the reference fuel cell.

R=4.7

5.4

8.5 mm

Zone:

1 2 3

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TABLE 2.1. INITIAL NUCLIDE DENSITIES IN THE CELL (atoms/cm³)

average in cell zone 1 zone 2 zone 3 Th-232 6.45E+21 2.11E+22

Pu-238 2.97E+18 9.72E+18 Pu-239 1.83E+20 5.99E+20 Pu-240 7.10E+19 2.32E+20 Pu-241 2.35E+19 7.69E+19 Pu-242 1.46E+19 4.78E+19

Cr 1.99E+20 8.14E+19 3.20E+20

Mn 1.26E+19 2.11E+19

Fe 5.20E+20 1.60E+20 8.46E+20

Ni 2.24E+20 3.76E+20

Zr 4.27E+21 4.37E+22

C 1.60E+18 2.68E+18

H 2.86E+22 4.80E+22

O 2.78E+22 4.41E+22 2.40E+22

The task to be performed for this benchmark exercise was defined as follows:

Calculate the fuel burnup at constant power (211 W/cm) as a function of time, not using any neutron poison for reactivity control. For a burnup of 0, 30, 40, and 60 MWd per kg of heavy metal report the following items:

(1) Neutron multiplication (keff);

(2) Total neutron flux;

(3) Average energy per fission;

(4) Residual amount of plutonium;

(5) Fraction of fissile plutonium;

(6) Amount of generated minor actinides;

(7) Average, (1-group, for the comparison) microscopic cross-sections for absorption, and fission for the heavy metal isotopes from ²³²Th through ²⁴⁴Cm.

2.1.1.2. Benchmark results

The comparison of the results achieved by the participants is displayed in Figs 2.2 - 2.8 and in Tables 2.2 and 2.3. The results show some deviations, e.g., in the calculated cell reactivity (ranging from ∆ρ ≈ 2% initially to ∆ρ ≈ 5% at the end of burnup) and in the average effective energy per fission of the respective mixture of fissionable isotopes (discrepancy up to 4%).

The results for the incineration rate of the plutonium isotopes, and for the buildup of minor actinides out of plutonium, as well as the ²³³U buildup from ²³²Th are in a good agreement.

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Based on these results, the participants of the CRP came to the conclusion that:

ದ generally, the different methods and databases are comparable to the degree, needed to permit sharing of the research for different reactor types among different groups of countries;

ದ however, a second benchmark should be performed for the special heterogeneity of a PWR-lattice.

0.70 0.80 0.90 1.00 1.10 1.20

0 20 40 60

BU [MWd/kg]

K

Germany Russia China Korea India USA Japan Netherlands

FIG. 2.2. Neutron multiplication vs. heavy metal burnup.

Total Neutron Flux

0.00 0.50 1.00 1.50 2.00 2.50 3.00 3.50 4.00 4.50

0.00 10.00 20.00 30.00 40.00 50.00 60.00

BU [MWd/kg]

n/10**14/(cm**2 sec)

FIG. 2.3. Total neutron flux vs. heavy metal burnup.

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190 195 200 205 210 215

0 20 40 60

BU [MWd/ kg]

MeV

Germany Russia China Korea India J^apan USA Netherlands

FIG. 2.4. Average energy per fission vs. heavy metal burnup.

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0

0 10 20 30 40 50 60

FIG. 2.5. (Pu/Pu_initial) vs. heavy metal burnup (MWd/t).

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0.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80

0 10 20 30 40 50 60

Burnup (MWd / kg)

FIG. 2.6. (Pu-fiss/Pu-total) vs. heavy metal burnup.

0 0.01 0.02 0.03 0.04 0.05 0.06 0.07 0.08 0.09

0 30 40 60

BU ( MWd / kg )

Germany Russia Korea USA Japan Netherlands

0.0 0.1 0.2 0.3 0.4 0.5 0.6

0 20 40 60

Burnup (MWd/kg)

FIG. 2.7. Minor actinides/initial plutonium. FIG. 2.8. ²³³U bred from Th/initial fissile plutonium.

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TABLE 2.2. CROSS-SECTIONS AT BURNUP = 0 MWd/kg

Germany Russia China Korea India USA Japan NL

Th-232

fission 0.0249 0.0260 0.0247 0.0281 0.0259 0.0276 0.029 0.0264 absorption 0.7889 0.8579 0.7986 0.8557 0.8664 0.858 0.817 0.849 Pa-233

fission 0.1631 0.3879 0.175 0.3836 0.128 0.1743 absorption 24.95 28.11 24.55 27.7 24.97 25.11 U-233

fission 0.531 0.4961 0.5607 0.549 0.565 absorption 22.46 21.97 22.66 20.9 22.22 U-235

fission 0.3466 0.2119 0.3528 0.352 0.357 absorption 10.85 9.669 10.11 10.12 10.6 U-238

fission 0.11 0.1062 0.114 0.114 absorption 8.219 8.200 7.785 7.923 Np-237

fission 0.528 0.5607 0.5656 0.571 0.564 absorption 27.92 25.56 27.14 27.21 27.7 Np-239

fission 0.621 0.6383 0.6578 0.659 0.664 absorption 15.28 15.03 15.9 14.55 14.74 Pu-238

fission 1.981 2.171 1.963 2.04 2.037 2.038 2.05 absorption 17.94 17.60 17.3 17.5 17.05 16.83 17.3 Pu-239

fission 44.63 44.98 44.31 43.23 44.1 44.05 43.89 44.09 absorption 69.74 69.90 69.00 66.98 68.5 69.06 68.11 68.4 Pu-240

fission 0.4502 0.4448 0.4455 0.4754 0.4988 0.475 0.4979 absorption 33.91 19.84 35.49 19.14 27.2 29.68 22.81 23.63 Am-241

fission 0.8569 0.8580 0.9062 0.923 0.881 0.9617 absorption 62.47 63.68 60.6 65.09 60.24 64.72 Am-242m

fission 289.3 312.0 277.31 264.0 287.1 absorption 346.0 382.9 331.8 314.9 352.1 Am-243

fission 0.4479 0.4638 0.5038 0.485 absorption 49.67 49.51 48.77 50.04 Cm-242

fission 0.4503 1.295 0.4705 1.312 1.0818 absorption 5.699 5.326 4.222 5.056 4.952 Cm-243

fission 0.9772 0.8405 1.082 0.872 1.041 absorption 18.29 19.6 23.55 19.90 18.01

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TABLE 2.3. CROSS-SECTIONS AT BURNUP = 60 MWd/kg

Germany Russia China Korea India USA Japan NL

Th-232

fission 0.0228 0.0240 0.0226 0.0246 0.0237 0.0254 0.025 0.0234 absorption 1.0893 1.141 1.085 1.295 1.170 1.1425 1.163 1.132 Pa-233

fission 0.1492 0.3568 0.1482 0.1512 0.3500 0.3669 0.113 0.155 absorption 21.78 23.85 21.78 22.96 24.3 23.6 22.21 21.34 U-233

fission 0.5056 0.4561 0.5025 0.5093 0.468 0.483 0.528 absorption 23.56 16.74 24.29 24.44 16.73 19.66 19.1 U-235

fission 0,3086 0.1951 0.3116 0.2818 0.203 0.309 0.309 absorption 9.252 8.289 9.627 6.829 8.342 8.82 8.64 U-238

fission 0.1006 0.0792 0.1071 0.10 0.101 absorption 7.595 7.227 7.207 7.241 7.234 Np-237

fission 0.4830 0.5160 0.4796 0.483 0.5083 0.505 0.509 absorption 34.69 31.66 34.07 39.19 34.59 35.48 34.19 Np-239

fission 0.5667 0.5864 0.5581 0.5981 0.583 0.6001 absorption 15.69 15.08 19.68 15.03 15.25 15.01 Pu-238

fission 2.468 2.635 2.655 2.58 2.538 2.615 2.525 absorption 38.75 36.92 47.22 38.9 36.96 39.61 36.36 Pu-239

fission 0.4115 0.4089 0.4087 0.4060 1.437 0.419 0.446 absorption 29.49 16.75 31.91 14.00 21.70 79.95 17.24 17.84 Am-241

fission 1.090 1.118 1.269 1.283 1.285 1.204 1.256 absorption 111.8 116.3 136.5 123.9 119.6 126.4 116.01 Am-242m

fission 718.7 784.7 888.5 745.8 706.01 absorption 864.0 967.6 1069 892.2 870.5 Am-243

fission 0.4094 0.4262 0.4363 0.4268 0.431 0.423 absorption 42.02 41.98 46.39 43.09 44.22 41.73 Cm-242

fission 0.9174 0.8148 0.9747 0.829 0.964

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2.1.2. Lattice calculations for LWR

While for a Pebble-bed HTR, having a nearly homogeneous core structure, a neutronics calculation for the entire core is regarded to be the adequate step following the cell calculation, an additional inter-comparison of the heterogeneous lattice calculation methodology appeared to be useful in case of the PWR. Thus, for the PWR part of the CRP, a second benchmark was established. Five countries participated in this benchmark: India, Israel, Japan, Republic of Korea, and Russian Federation. The benchmark was designed to compare assembly-level calculation methods, by defining a 2-D lattice simulating a typical PWR fuel assembly.

2.1.2.1. Benchmark definition and tasks General:

17x17 array of the fuel rods, including 25 water hole positions.

No guide tubes material. No assembly casing.

No buckling. Quarter assembly symmetry.

Burnup calculations with constant specific power of 37.7 MW/t (initial heavy metal).

Geometry:

Outer dimensions, cm: 22.662×22.662 Cell pitch, cm: 1.33306 Fuel pellet radius, cm: 0.4127 Cladding thickness, cm: 0.0617 Equiv. cell radius, cm: 0.7521

Material compositions (atoms/barn x cm):

Fuel:

5% PuO2 + 95% ThO2. Temperature: 900 K.

Th-232 2.0592E-2 Pu-238 2.2900E-5 Pu-239 7.4780E-4 Pu-240 2.9030E-4 Pu-241 1.5340E-4 Pu-242 5.0100E-5 O-16 4.3710E-2 Cladding:

Natural Zr. Temperature: 600 K.

Zr-nat. 4.3241E-2 Moderator:

Light water, with 500 ppm natural boron. Temperature: 573 K.

H-1 4.7708E-2 O-16 2.3854E-2 B-10 3.9518E-6 B-11 1.5906E-5

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Results for comparison:

(1) Criticality as a function of burnup. Burnup range from 0 to 60 GWd/t;

(2) Fuel composition as a function of burnup (major actinides and fission products);

(3) Local pin-by-pin power distribution;

(4) Moderator temperature coefficient for 0 and 60 GWd/t;

(5) Doppler coefficient for 0 and 60 GWd/t;

(6) Soluble boron worth for 0 and 60 GWd/t.

2.1.2.2. Benchmark results

Infinite multiplication factor:

The results are summarized in Table 2.4 and Fig. 2.9.

TABLE 2.4. k_inf AS A FUNCTION OF BURNUP Burnup,

GWd/T

Russian

Federation Japan Republic of

Korea India Israel 0 1.189 1.1987 1.1864 1.2076 1.1956 0.5 1.1569 1.1670 1.1551 1.1736 1.1643

20 1.0298 1.0521 1.0303 1.0372 1.0290 40 0.9147 0.9527 0.9167 0.9104 0.9119 60 0.8315 0.8657 0.8310 0.8294 0.8314 Significant discrepancies are found between kinf values: ~ 2.3% ∆K at BOL and ~ 3.5% ∆K at EOL. It is also noted that there is no clear burnup dependency of the discrepancies.

0.8 0.9 1.0 1.1 1.2

0 10 20 30 40 50 60

Russia Japan Korea India Israel

Fig. 2.9. k_inf as a function of burnup.

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Fuel composition:

The results obtained for the composition of the actinide fuel are listed in Table 2.5 as a function of the heavy metal burnup. Note that there is:

ದ a good agreement for the ²³⁸Pu concentration;

ದ a reasonable agreement for the even plutonium isotopes concentrations;

ದ significant discrepancies are found for the odd plutonium isotopes;

ದ a reasonable agreement in the case of the ²³²Th and ²³³U concentrations.

TABLE 2.5. FUEL COMPOSITION (ACTINIDES) AS A FUNCTION OF BURNUP

Number Density (atoms/barn×cm)

Burnup, GWd/T Russia Japan Rep. of Korea India Israel Th-232

0.0 2.059×10^-2 2.059×10^-2 2.059×10^-2 2.059×10^-2 2.059×10^-2

0.5 2.059×10^-2 2.059×10^-2 2.059×10^-2 - - 20.0 2.037×10^-2 2.036×10^-2 2.037×10^-2 2.036×10^-2 2.037×10^-2

40.0 2.011×10^-2 2.008×10^-2 2.011×10^-2 - 2.010×10^-2 60.0 1.977×10^-2 1.975×10^-2 1.978×10^-2 1.970×10^-2 1.977×10^-2

Pu-238

0.0 2.290×10^-5 2.290×10^-5 2.290×10^-5 2.290×10^-5 2.290×10^-5

0.5 2.279×10^-5 2.279×10^-5 2.279×10^-5 - - 20.0 1.940×10^-5 1.952×10^-5 1.928×10^-5 1.829×10^-5 1.937×10^-5

40.0 1.834×10^-5 1.879×10^-5 1.798×10^-5 - 1.793×10^-5 60.0 1.687×10^-5 1.816×10^-5 1.636×10^-5 7.488×10^-6 1.611×10^-5

Pu-239

0.0 7.478×10^-4 7.478×10^-4 7.478×10^-4 7.478×10^-4 7.478×10^-4

0.5 7.348×10^-4 7.351×10^-4 7.349×10^-4 - - 20.0 3.174×10^-4 3.270×10^-4 3.175×10^-4 2.993×10^-4 3.147×10^-4

40.0 0.810×10^-4 0.961×10^-4 0.820×10^-4 - 0.773×10^-4 60.0 0.118×10^-4 0.170×10^-4 0.121×10^-4 0.479×10^-4 0.105×10^-4

Pu-240

0.0 2.903×10^-4 2.903×10^-4 2.903×10^-4 2.903×10^-4 2.903×10^-4

0.5 2.911×10^-4 2.909×10^-4 2.911×10^-4 - - 20.0 2.826×10^-4 2.678×10^-4 2.820×10^-4 2.846×10^-4 2.853×10^-4

40.0 1.981×10^-4 1.845×10^-4 1.991×10^-4 - 2.014×10^-4 60.0 0.809×10^-4 0.839×10^-4 0.874×10^-4 0.670×10^-4 0.846×10^-4

Pu-241

0.0 1.534×10^-4 1.534×10^-4 1.534×10^-4 1.534×10^-4 1.534×10^-4

0.5 1.540×10^-4 1.543×10^-4 1.541×10^-4 - - 20.0 1.591×10^-4 1.703×10^-4 1.605×10^-4 1.545×10^-4 1.578×10^-4

40.0 1.233×10^-4 1.360×10^-4 1.231×10^-4 - 1.214×10^-4 60.0 0.650×10^-4 0.741×10^-4 0.641×10^-4 0.539×10^-4 0.639×10^-4

Pu-242

0.0 0.5010×10^-4 0.5010×10^-4 0.5010×10^-4 0.5010×10^-4 0.5010×10^-4

0.5 0.5050×10^-4 0.5043×10^-4 0.5051×10^-4 - - 20.0 0.7088×10^-4 0.6813×10^-4 0.7248×10^-4 0.7203×10^-4 0.7020×10^-4

40.0 0.9877×10^-4 0.9245×10^-4 1.0380×10^-4 - 0.9832×10^-4 60.0 1.1890×10^-4 1.1030×10^-4 1.2880×10^-4 1.1624×10^-4 1.1940×10^-4

U-233

0.0 - - - - - 0.5 0.7319×10^-6 0.7918×10^-6 0.7378×10^-6 - - 20.0 1.5150×10^-4 1.5996×10^-4 1.5350×10^-4 1.5960×10^-4 1.5330×10^-4

40.0 2.6120×10^-4 2.7492×10^-4 2.6400×10^-4 - 2.6750×10^-4 60.0 3.1350×10^-4 3.3109×10^-4 3.1600×10^-4 3.1910×10^-4 3.2350×10^-4

U-234

0.0 - - - - - 0.5 0.2361×10^-7 0.2522×10^-7 0.1556×10^-7 - - 20.0 0.8565×10^-4 0.9714×10^-5 0.8025×10^-5 0.9627×10^-5 0.7913×10^-5

40.0 2.6680×10^-4 2.8855×10^-5 2.5200×10^-5 - 2.5290×10^-5 60.0 5.3200×10^-4 5.4315×10^-5 4.9070×10^-5 6.1950×10^-5 5.0450×10^-5

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Pin-by-pin power distribution:

The local power distributions are shown in Figs 2.10a and 2.10b for the BOL and 60 GWd/t burnup points, respectively. A very good agreement for the BOL is indicated, with less than 2% relative power differences. The “hot rod” is identified with almost identical power of 1.124. A divergence of the local power values with burnup resulted in 5–10% differences, which may be partially attributed to different fissile (odd numbers) plutonium isotope concentrations.

W

1.071 1.025 xxxx Russia

1.068 1.019 xxxx Japan 0 GWd / t

1.071 1.015 xxxx Korea 1.075 1.015 xxxx Israel

1.059 1.014 1.015

1.068 1.020 1.021

1.071 1.015 1.017

1.076 1.016 1.018

1.068 1.072

1.068 1.071

1.070 1.073

W

1.075 1.078 W

1.063 1.016 1.009 1.084 1.080 1.066 1.018 1.023 1.084 1.068 1.070 1.015 1.019 1.087 1.077 1.074 1.015 1.020 1.091 1.074 1.072 1.013 1.005 1.08 1.122 1.061 1.014 1.019 1.085 1.107 1.066 1.011 1.016 1.088 1.126 1.069 1.011 1.017 1.092 1.124

W

1.053 1.058 1.099 1.049 0.963 1.051 1.054 1.081 1.045 0.965 1.056 1.061 1.100 1.053 0.964 W

1.060 1.064 W

1.100 1.055 0.964

1.051 0.979 0.98 1.033 0.973 0.922 0.910 0.886 1.021 0.977 0.978 1.024 0.976 0.941 0.915 0.899 1.031 0.978 0.980 1.034 0.976 0.928 0.902 0.884 1.034 0.980 0.981 1.036 0.976 0.931 0.904 0.884

0.925 0.924 0.921 0.930 0.922 0.907 0.889 0.889 0.876 0.929 0.925 0.925 0.929 0.921 0.910 0.901 0.896 0.898 0.926 0.923 0.922 0.924 0.914 0.898 0.886 0.878 0.876 0.926 0.922 0.921 0.923 0.914 0.897 0.885 0.876 0.872

FIG. 2.10a. Relative power distribution.

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W

0.951 0.926 xxxx Russia

1.037 1.027 xxxx Japan 60 GWd / t

1.025 1.015 xxxx Korea 1.054 1.027 xxxx Israel

0.944 0.931 0.939

1.037 1.027 1.027

1.025 1.015 1.015

1.055 1.027 1.028

0.955 0.96

1.037 1.038

1.024 1.025

W

1.054 1.056 W

0.959 0.952 0.950 0.986 0.975 1.034 1.024 1.027 1.042 1.039 1.023 1.013 1.014 1.027 1.022 1.051 1.025 1.028 1.064 1.057 0.979 0.957 0.967 1.001 0.922 1.028 1.018 1.021 1.038 1.041 1.019 1.010 1.011 1.024 1.026 1.046 1.019 1.023 1.060 1.074

W

0.984 0.991 1.024 1.003 1.014 1.017 1.019 1.024 1.007 0.978 1.013 1.013 1.015 1.004 0.983 W

1.034 1.037 W

1.052 1.022 0.969

1.026 1.009 1.022 1.032 1.030 1.051 1.029 1.027 0.997 0.989 0.989 0.997 0.985 0.971 0.957 0.946 1.003 0.994 0.994 1.001 0.989 0.978 0.968 0.960 1.013 0.988 0.989 1.012 0.981 0.952 0.930 0.912

1.054 1.045 1.049 1.055 1.044 1.036 1.024 1.017 1.013 0.966 0.965 0.965 0.965 0.961 0.954 0.948 0.944 0.945 0.985 0.983 0.983 0.982 0.977 0.970 0.962 0.956 0.953 0.955 0.952 0.952 0.952 0.942 0.927 0.914 0.903 0.897

FIG. 2.10b. Relative power distribution.

Temperature coefficients and boron worth values are presented in Table 2.6.

TABLE 2.6. TEMPERATURE COEFFICIENTS AND BORON WORTH (×10^–4)

0 GWd/T 60 GWd/t

MTC^a DC^b BW^c MTC DC BW

Russian

Federation -3.500 -0.280 -0.380 -1.5 -0.360 -1.100 Japan -2.696 -0.283 -0.341 -0.969 -0.378 -0.864 Republic of

Korea -3.200 -0.311 -0.408 -1.289 -0.397 -1.125 Israel -3.333 -0.292 -0.400 -1.142 -0.477 -1.119

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Tm

K K MTC K

∆

∗

= ∆

2 1

- Moderator temperature coefficient, Tm – moderator temperature.

f 2

1 K T

K DC K

∆

∗

= ∆ - Doppler coefficient, Tf – fuel temperature.

C K K BW K

2

1∗ ∗∆

= ∆ - Soluble boron worth, C – boron concentration in ppm.

Note 1: All temperature coefficients are negative, the burnup dependence, i.e., plutonium depletion effect is correct.

Note 2: All BOL values show reasonable agreement, the divergence of the EOL values may be attributed to different plutonium concentrations.

2.2. EVALUATION OF THE POTENTIAL OF LWRs, HTRs, HWRs, AND MSRs FOR PLUTONIUM INCINERATION

2.2.1. Incentives

The aim of the research during the second stage of the CRP was to find fuelling strategies, which – on the basis of proven reactor technology are suitable to incinerate plutonium most effectively on the one hand, and to minimize the amount of plutonium to be disposed, on the other hand. Only plutonium of the first generation, namely typical LWR-plutonium, and weapons plutonium were regarded within scope of this CRP.

Four types of reactors were investigated in view of their potential to burn plutonium, each by one group of countries. Israel, Republic of Korea, Russian Federation, and the USA have done research on LWRs; China, the Netherlands and Germany have studied plutonium burning in Modular HTRs; India studied the respective potential of the PHWR, and Japan that of the MSR.

Two characteristic values – aiming at two different optimization goals — may describe the effectiveness of plutonium incineration in the different reactors:

(1) The amount of plutonium, which is burned per unit of produced electric energy.

Maximization of this characteristic optimizes the reduction rate of existing plutonium stockpiles.

(2) The relation between the amount of plutonium, which is burned during the lifetime of the fuel elements, and the amount of plutonium, which is residual in the unloaded fuel.

Maximization of this characteristic minimizes the plutonium quantity, which either has to be finally disposed or has to be re-fabricated a second time.

Some initial remarks have to be made concerning the admissibility of a comparison and the assessment of the data presented by the countries participating in this CRP.

Each country did the research on its favorite reactor concept, using its own methods and computer codes as well as its specific database. Although, as already mentioned, benchmark calculations for a PWR-cell and for a PWR-lattice were performed during the first stage of this CRP, the numerical simulation of a complete reactor and of his fuel cycle is a much more complex matter, and the results may well be influenced by the degree of detail of the

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numerical modeling. Also, the restrictions of the reactor fuelling in view of safety aspects (power peaking, temperature coefficients, transient behavior) may have been observed by the various groups up to different degrees. Nevertheless, the information resulting from the comparison presented in this report seems to be well appropriate and sufficiently reliable to show the potential of the different reactor concepts incinerating plutonium in an effective manner, and to help the suggesting a demonstration of plutonium burning in a reactor in one of the member countries.

By the time the data presented here was compiled the Netherlands had not yet performed calculations for an entire reactor, but for an HTR-fuel cell only. Thus, general effects and operational constraints resulting from, e.g., neutron leakage, local power peaking, and requirements of reactor control may not be observed. Furthermore, a fuel cell containing only plutonium and no thorium was investigated. Therefore, the data is excluded from the summary tables in order to avoid confusion. Results of this research have been published elsewhere [1].

2.2.2. Results

Tables 2.7 and 2.8 show the results of the research work, which has been done based on LWR plutonium and on weapons grade plutonium, respectively. Blank spaces indicate that results have not yet been evaluated so far. The thermal efficiency 0.33 for water cooled reactors and a value 0.4 for HTR and for MSR has been assumed when compiling these tables. Two countries (Russian Federation and Germany) investigated 2 different alternative fuelling strategies each, to be used for their favored reactor concept:

Russian Federation studied the WWER-reactor applying:

a) partial (1/3 of the core) loading with PuO2 – ThO2 fuel (“Partial Pu-Inv.”);

b) loading PuO2-ThO2 in the entire core (“Full Pu-Inv.”).

Germany in its study made use of the capability of the coated particle fuel of the HTR for a very high heavy metal burnup in order to minimize the amount of residual plutonium in the discharged fuel elements (core “minimal residual plutonium”). The possibility to increase the incineration rate by a reduction of the burnup is demonstrated by the case “increased inciner.”

(for “LWR-Pu” only, Table 2.7).

The amount of LWR plutonium (Table 2.7), which is burned per unit of produced energy in the different reactor concepts generally decreases with increasing heavy metal burnup of the fuel. In view of the minimization of the residual plutonium in the discharged fuel elements, however, a distinct advantage of the high burnup is obvious. This is indicated by the ratio between the amount of plutonium, which is burned until the fuel element is discharged, and the residual amount, which either has to be disposed or has to be re-fabricated a second time (ratio Pu-burned/Pu-discharged).

Incinerating weapons grade plutonium (Table 2.8) is generally more effective, especially if a high burnup is applied as for PHWR (India) and HTR (China and Germany). The practically complete absence of plutonium isotopes higher than ²⁴⁰Pu in the fresh fuel shifts the occurrence of increased parasitic neutron absorption by ²⁴²Pu and by minor actinides to a high burnup of this fuel. Thus, the plutonium can be burned to a high degree, achieving a high yearly destruction rate at the same time. The ratio between burnt and residual plutonium can go up to a value of 5.9 compared to 4.2 in the case of LWR plutonium.

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Figure 2.11a and b expressively elucidates these relations. The incineration rate of plutonium (ordinate) is given in units of 200 kg/GWela, which is approximately the amount of plutonium produced by current LWRs. In other words, this number indicates, how many LWRs units could have their spent fuel plutonium incinerated by one unit of the considered plutonium burner.

2.2.3. Conclusions

ದ Incinerating LWR plutonium:

Water cooled reactors (LWR and HWR), having a relatively low heavy metal burnup (40-50 GWd/to) reach a large plutonium-incineration rate in the range of about 700-850 kg/GWela. On the other hand, this amount of incinerated plutonium is small compared to that remaining for disposal or reuse at the end of burnup (Pu burned/Pu discharged equals 0.8-1.7, which corresponds to a residual plutonium fraction in the range of 56 to 37%). Achieving a large heavy metal burnup (e.g., by HTR-fuel) results in a smaller amount of incinerated plutonium per unit of produced energy (500-650 kg/GWela), but distinctly reduces the fraction of residual plutonium down to 19%.

ದ Burning weapons grade plutonium:

The higher neutronic value of weapons grade plutonium and the strongly reduced build-up of minor actinides out of ²⁴²Pu generally make weapons grade plutonium incineration more effective than for LWR plutonium. While the amount of incinerated weapons plutonium per unit of produced electric energy is comparable to the incineration rate of LWR plutonium, the quantity of residual plutonium can be strongly reduced in case of weapons grade plutonium. The ratio “Pu-burned/Pu-discharged”

equals 1.5 to 2 (40 to 33% residual plutonium fraction) for LWRs, about 4 (20% residual plutonium fraction) for the HWR and up to 6 (14% residual plutonium fraction) in case of the HTR. Here, the advantage of reactors having high burnups, becomes obvious once more.

TABLE 2.7. BURNING LWR PLUTONIUM: MASS BALANCE: kg/GW_ela (FULL POWER)

China Germany (HTR) India

(PHWR) Israel +USA (LWR)

Japan (MSR )

Republic of Korea (LWR)

Russian Federation (LWR)

minimal

resid – plutonium

increased

inciner. Partial Pu-Inv. Full

Pu-Inv.

U-235/

U-233 charged

- 624/0 578/0 6/0 612 / 0

Pu-charged 2521 615 929 1098 1419 1435 1708 519 1803 Pu-discharged 1576 119 288 405 614 435 875 401 953 Pu-burned 945 496 641 693 805 1000 833 117 850 Ratio

Pu-burned/

Pu-discharged

0.6 4.2 2.2 1.7 1.3 2.3 0.95 0.29 0.89

U-233 produced

116 161 286 141 366 100 291

Average HM burnup (MWd/kg)

192 128 46 100 40 41 40

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TABLE 2.8. BURNING WEAPONS GRADE PLUTONIUM: MASS BALANCE: kg/GWela(FULL POWER)

China (HTR)

(Case

“11g”)

Germany (HTR) India

(PHWR) Israel + USA (LWR)

Japan

(MSR) Republic of Korea (LWR)

Russian Federation (LWR)

Partial Pu-Inv.

Full Pu-Inv.

U-235/

U-233 charged

- 188 - - 7 612 -

Pu-charged 1097 820 725 1095 1425 1264 354 1220 Pu-discharged 212 118 141 361 521 507 266 462 Pu-burned 885 702 584 734 904 757 87 758 Ratio

Pu-burned/

Pu-discharged

4.2 5.9 4.1 2.0 1.7 1.5 0.33 1.64

U-233 produced 207 151 204 233 397 100 294 Average HM

burnup (MWd/kg)

103 128 70 100 40 41 41

LWR-plutonium

0 1 2 3 4 5 6

Israel+USA Russia, FI Korea India China Germany, incr. Inc.

Germ., m.r.Pu

Japan

Pu-burned / 200 kg / GWa-el Pu burned / Pu discharged

FIG. 2.11a. Burning LWR-grade plutonium.

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Weapons-grade plutonium

0 1 2 3 4 5 6

Israel+USA Russia,FI Korea India China G erman y Japan

Pu-burned / 200 kg / GWa-el Pu burned / discharged

FIG. 2.11b. Burning weapons grade plutonium.

2.3. EFFECT OF PLUTONIUM INCINERATION ON THE TOXICITY OF DISPOSED NUCLEAR WASTE

2.3.1. Incentives and database

The research reported in Section 2.2 primarily aims at the minimization of the proliferation risk by minimizing the plutonium production and maximizing the plutonium incineration. The question still remains, whether and to which degree the incineration of plutonium furthermore is an appropriate tool to significantly reduce the hazard potential of the nuclear waste, which in the end remains for final disposal.

The most common procedure in order to assess the toxicity of nuclear materials is based on the recommendations of the International Commission on Radiological Protection, defining

“Annual Limits on Intake” for the radio-toxic isotopes. Recent recommendations are given in Sv/Bq and are called “Dose Coefficients for Intake (DCI)”. It was agreed to use the values according to ICRP Publications 68 /ICRP 1994/ and 61 /ICRP 1991/ as the common database within the frame of this CRP. A comparison between the waste of uranium-fuelled LWRs (providing no reprocessing of the discharged fuel) on the one hand and the waste produced at a scenario applying plutonium burning reactors on the other hand helps to assess the related effect on the toxicity.

2.3.2. Toxicity benchmark

In order to assure that the computer codes used in the different countries have been correctly updated in the sense of the statements of Section 2.3.1, the first step of this evaluation was a benchmark with respect to the toxicity of the spent fuel resulting from one year operation of a 1GWel reference-PWR. The composition of the unloaded heavy metal isotopes is defined in

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TABLE 2.9. DISCHARGE RATE OF HEAVY METAL ISOTOPES (kg/year) FOR A TYPICAL PWR*

U-234 4.51E 00

U-235 2.70E 02

U-236 1.07E 02

U-237 2.70E-01

U-238 2.69E 04

Np-237 1.11E 01

Np-239 2.27E 00

Pu-238 3.12E 00

Pu-239 1.43E 02

Pu-240 5.78E 01

Pu-241 3.11E 01

Pu-242 1.02E 01

Am-241 8.87E-01 Am-242m 1.66E-02 Am-242 2.12E-03

Am-243 1.77E 00

Am-244 6.41E-05 Cm-242 2.38E-01 Cm-243 5.17E-03 Cm-244 4.56E-01 Cm-245 1.66E-02

Ȉ 2.75E 04

*All data normalized to 1000 MWel the electric power output and 300 full-power days.

The tasks to be commonly performed then were to evaluate:

ದ The ingestion hazard of the complete heavy metal waste;

ದ The inhalation hazard of the complete heavy metal waste;

ದ The ingestion hazard of the heavy metal waste remaining after separation of 99% of all plutonium isotopes;

ದ The inhalation hazard of the heavy metal waste remaining after separation of 99% of all plutonium isotopes.

Dose coefficients of intake (DCI) for the heavy metal isotopes and for the fission products, to be commonly used, are given in Table 2.10.

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TABLE 2.10. DOSE COEFFICIENTS OF INTAKE (DCI) Unit: Sv/Bq

NUCL: Isotope identification number = Z×10000 + W×10 + IS, with Z: the atomic number

W: the atomic weight

IS: equal 0 or 1 for ground or metastable state, respectively DCI-W: DCI value for water

DCI-A: DCI value for air References:

ICRP Publication 68 (1994)

ICRP Publication 61 (1991), DCI calculated for reference dosis 20 mSv/a

NUCL DCI-W DCI-A NUCL DCI-W DCI-A NUCL DCI-W DCI-A

20040 0.E+00 0.E+00 882250 1.E-07 6.E-06 942360 9.E-08 2.E-05 812070 5.E-10 5.E-09 882260 3.E-07 2.E-05 942380 2.E-07 4.E-05 812080 5.E-10 5.E-09 882280 7.E-07 3.E-06 942390 3.E-07 5.E-05 812090 5.E-10 5.E-09 892250 2.E-08 8.E-06 942400 3.E-07 5.E-05 822060 0.E+00 0.E+00 892270 1.E-06 6.E-04 942410 5.E-09 9.E-07 822070 0.E+00 0.E+00 892280 4.E-10 3.E-08 942420 2.E-07 4.E-05 822080 0.E+00 0.E+00 902270 9.E-09 1.E-05 942430 9.E-11 1.E-10 822090 6.E-11 3.E-11 902280 7.E-08 4.E-05 942440 2.E-07 4.E-05 822100 7.E-07 1.E-06 902290 5.E-07 1.E-04 942450 7.E-10 7.E-10 822110 2.E-10 6.E-09 902300 2.E-07 4.E-05 952410 2.E-07 4.E-05 822120 6.E-09 3.E-08 902310 3.E-10 4.E-10 952421 2.E-07 4.E-05 822140 1.E-10 5.E-09 902320 2.E-07 4.E-05 952420 3.E-10 2.E-08 832090 0.E+00 0.E+00 902330 5.E-10 5.E-09 952430 2.E-07 4.E-05 832100 1.E-09 8.E-08 902340 3.E-09 7.E-09 952440 5.E-10 2.E-09 832110 9.E-10 3.E-08 912310 7.E-07 1.E-04 952450 6.E-11 8.E-11 832120 3.E-10 4.E-08 912320 7.E-10 1.E-08 962420 1.E-08 5.E-06 832130 2.E-10 4.E-08 912330 9.E-10 4.E-09 962430 2.E-07 3.E-05 832140 1.E-10 2.E-08 912341 5.E-10 5.E-09 962440 1.E-07 3.E-05 842100 2.E-07 3.E-06 912340 5.E-10 6.E-10 962450 2.E-07 4.E-05 842110 9.E-10 3.E-08 922320 3.E-07 4.E-05 962460 2.E-07 4.E-05 842120 9.E-10 3.E-08 922330 5.E-08 9.E-06 962470 2.E-07 4.E-05 842130 9.E-10 3.E-08 922340 5.E-08 9.E-06 962480 8.E-07 1.E-04 842140 9.E-10 3.E-08 922350 5.E-08 8.E-06 962490 3.E-11 5.E-11 842150 9.E-10 3.E-08 922360 5.E-08 8.E-06 962500 4.E-06 8.E-04 842160 9.E-10 3.E-08 922370 8.E-10 2.E-09 972490 1.E-09 2.E-07 842180 9.E-10 3.E-08 922380 4.E-08 7.E-06 972500 1.E-10 1.E-09 852170 9.E-10 3.E-08 922390 3.E-11 4.E-11 982490 4.E-07 7.E-05 862190 3.E-11 7.E-11 922400 1.E-09 8.E-10 982500 2.E-07 3.E-05 862200 3.E-11 7.E-11 932360 2.E-08 3.E-06 982510 4.E-07 7.E-05 862220 3.E-11 7.E-12 932370 1.E-07 2.E-05 982520 9.E-08 2.E-05 872210 9.E-10 3.E-08 932380 9.E-10 2.E-09 982530 1.E-09 1.E-06 872230 2.E-09 1.E-09 932390 8.E-10 1.E-09 982540 4.E-07 4.E-05 882230 1.E-07 7.E-06 932401 5.E-10 5.E-09 992530 6.E-09 3.E-06 882240 7.E-08 3.E-06 932400 8.E-11 1.E-10